1. FIELD OF THE INVENTION
This invention relates to vent assembly systems and, more particularly, to a vent system for the release of vapor pressure in the core support cylinder of a nuclear reactor pressure vessel.
2. DESCRIPTION OF THE PRIOR ART
The conventional nuclear reactor pressure vessel generally comprises a longitudinally disposed cylindrical structure, closed at both ends by a convex base and a domed roof, and having reactor coolant inlet and outlet nozzles disposed in angular separation in a plane transverse to the longitudinal axis of the vessel and protruding therethrough. Housed within the pressure vessel structure is, among others, the nuclear reactor core ordinarily supported by a core support cylinder or shell suspended from an annular flange formed on the inner surface of the vessel. The core support cylinder comprises a distribution hoop, from which the primary coolant discharges through the outlet nozzles, and a thermal shield-skirt assembly, which supports the fuel elements in the reactor core and which in conjunction with the distribution hoop and the internal wall of the reactor pressure vessel serves as an annular hydraulic guide for the primary inlet coolant.
In operation, the fluid coolant, in forced circulation, enters the pressure vessel through the inlet nozzles, flows through the annular hydraulic guide formed between the inner surface of the pressure vessel and the core support cylinder, and rises through the reactor core whereupon it is discharged from the vessel through the outlet nozzles.
Furthermore, from a safety posture, nuclear reactor systems are generally enclosed in substantially leaktight concrete or steel containment structures to prevent radioactive materials such as gaseous, vaporized, solid or dissolved fission products from escaping from the containment in the event of a reactor accident. One such failure of the reactor system, the loss of coolant accident, or LOCA as it is commonly called in the nuclear reactor industry, results in flashing of the high pressure primary fluid, which pressurizes the containment, and rapid vaporization and therefore pressurization of the residual primary fluid remaining within the reactor vessel at the time of the LOCA. Accordingly, various safety systems have been suggested to suppress the vapor pressure build-up in the containment and, also, to provide emergency core cooling or flooding to the reactor core itself. However, in the event of accidental or catastrophic failure of the reactor system, a rapid pressure build-up in the reactor vessel, on the order of 500 pounds per square inch (psi) differential between the hoop area and the annulus in approximately one one hundredth (0.01) of a second, may occur and subsequent pressurization may prevent the emergency core coolant or flooding systems from adequately flooding the hot core with a coolant fluid. For example, during a LOCA involving the primary inlet coolant line, flow is interrupted, preventing the primary coolant from entering the core. The hot nuclear reactor core, however, continues to produce energy in the form of heat. The coolant pressure within the reactor vessel rapidly decreases to the saturation pressure, at which point coolant vapor accumulates and increases the pressure in the distribution hoop area. Moreover, a typical nuclear reactor vessel requires, during normal operation, on the order of hundreds of thousands of gallons of coolant per minute to adequately cool the reactor core and, therefore, an interruption of the primary coolant flow will allow excessive heat accumulation in the core and produce excessive heat transfer to the residual coolant within the vessel. The excessive heat transfer to the residual coolant may result in sufficient overpressurization of the coolant in the shell, such that, the decay heat removal systems or emergency core cooling systems may be prevented from flowing into and adequately cooling the core due to the over pressure of the heated residual coolant. Therefore, the performance of the emergency core cooling or flooding systems may be nullified or at least reduced resulting in a potential build-up of reactor core decay heat and a possible melt down of the core.
A simple "heavy" flapper type valve responsive to differential pressure across the valve and which will automatically open as the pressure increases in the core support cylinder has been suggested as a possible core cylinder venting means. However, since the pressure rise within the core support cylinder during, for example, a LOCA occurs almost instantaneously to hundreds of psi, the valve will "explosively" open, accelerating outwardly towards the vessel's internal wall at speeds approaching that of the sonic velocity of the vapor. Furthermore, since the dimension of the annulus between the core support cylinder and the internal wall of the reactor pressure vessel is ordinarily limited, an "explosively" opened valve of sufficient valve plate opening size for adequate pressure release will contact the reactor vessel's internal wall with sufficient force to cause severe deformation of the valve. Furthermore, in order to simplify the analytical analysis of valve plate deformation it has been suggested to incorporate into the valve system a "heavy" boss protruding outwardly from the plate into the annulus. However, since the "heavy" boss will also "explosively" contact the vessel wall there is the possibility that the "heavy" boss and valve plate will locally over stress the wall in the vicinity of the contact point. In addition, the valve-wall contact force may also substantially deform the valve hinge pin, such that the valve, if subsequently closed by incoming emergency coolant, may not reopen in responsive to a successive relatively low differential pressure build-up across the core support cylinder.
Accordingly, there is a need to provide venting means to the distribution hoop of the core support cylinder of a reactor pressure vessel which, during an accident or failure of the reactor, will relieve the pressure build-up in the core support cylinder, will not impose excessive loads upon the reactor vessel wall and which, subsequent to the initial response to a pressure build-up in the core support cylinder during, for example, a LOCA, will remain functional in response to successive relatively low differential pressure build-ups.